ML20235J535
| ML20235J535 | |
| Person / Time | |
|---|---|
| Site: | Oconee, Arkansas Nuclear, Crystal River, Rancho Seco, Bellefonte, 05000000, Crane |
| Issue date: | 09/28/1987 |
| From: | Stello V NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | Nunn S SENATE |
| Shared Package | |
| ML20235J539 | List: |
| References | |
| 2.206, NUDOCS 8710020005 | |
| Download: ML20235J535 (10) | |
Text
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SEP 2 8. 50 The Honorable Sam Nunn United States Senate Washington, D.C.
20510-1001
Dear Senator Nunn:
I am responding to your letter dated August 3,1987, to the NRC Director of -
Congressional Affairs enclosing a letter from Dr. Thomas F. Gross of Savannah, Dr. Gross asked that you support the efforts of the Union of Concerned Georgia.
Scientists (UCS) regarding the petition UCS submitted to the Nuclear Regulatory Commission (NRC).
This petition was submitted on February 10, 1987 pursuant to 10 CFR 2.206.
It was then referred to the NRC Office of Nuclear Reactor Regulation for appropriate action, in accordance with usual Commission practice.
The petition requested the Commission to:
(1) require all utilities operating or building nuclear power reactors designed by Babcock & Wilcox (B&W) to modify their nuclear generating facilities to correct alleged safety deficiencies; (2) hold public hearings on the sufficiency of these modifications to correct the alleged deficiencies; and (3) revoke the operating license or construction permit of any utility that does not meet the proposed requirements that emerge from (1) and (2) above.
In addition, the petition requested that NRC immediately suspend the operating licenses and construction permits of any utility operating or building nuclear power reactors designed by B&W.
By letter dated March 13, 1987, the Director, Office of Nuclear Reactor Regulation, informed the Union of Concerned Scientists that the concerns expressed in the petition did not warrant suspension of the operating licenses and construction permits in question.
The NRC staff is continuing to review the petition, and a formal decision on the requests stated in the petition is scheduled to be issued in the near future.
The staff's decision will be subject to the Commission's review.
In his letter, Dr. Gross stated his opinion that " operation of the B&W plants poses unreasonable risks to public health and safety." In support of his position, Dr. Gross provided five statements which, he indicated, were excerpts from NRC documents.
He also raised issues regarding the likelihood of a major accident at a nuclear plant, the effects of need for power on licensing decisions, and the operation of the Oconee plants.
The first three of the statements attributed to the NRC were as follows:
The unique B&W design makes these plants much more sensitive to f ailures and abnormal cunditions than other pressurized water reactors.
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The Honorable Sam Nunn !
Despite the modifications made to plant equipment and operator i
training at B&W plants following the 1979 accident at Three i
Mile Island, the number and complexity of accidents and near-misses, which the NRC calls ' events', have not decreased as expected.
The post-TMI modifications have not been sufficient.
1 The dangerous, complex accidents which punctuate the operating
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history of the B&W-designed plants continue to occur.
This verifies that the B&W design's weaknesses have not been correc-ted.
Under different circumstances, the consequences of such accidents could be much more severe.
It is true that since the accident at Three Mile Island (TMI) in 1979, the NRC l
l has recognized that the B&W plants are more sensitive in their response to l
operational events than other pressurized water reactors.
It is also true that after TMI, even though efforts were significantly intensified to improve B&W plants, the number and complexity of problems were still of concern.
The I
significant events that occurred at the Davis-Besse and Rancho Seco nuclear plants in June and December of 1985 respectively raised the level of concern significantly.
Even though the events at Davis-Besse and Rancho Seco were significant, the root causes of the events were not generic in nature.
A decision to snut down the other B&W plants would have to be based on a safety problem common to all of the plants.
The event at Davis-Besse was the result of inattentiveness to
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the care of plant equipment, and the Rancho Seco event was caused by a design l
deficiency in the integrated control system (ICS) peculiar to Rancho Seco.
Because of differences in ICS design, no other B&W plants would have responded as Rancho Seco did to a loss of power to the ICS.
Thus, because the Davis-Besse i
l and Rancho Seco events were not generic in nature and because many post-TMI modifications had been implemented, with positive results, at other B&W plants, the staff concluded that shutdown of the other B&W plants was not warranted.
However, given the level of concern that existed concerning the B&W plant design, the NRC Executive Director for Operations directed in January 1986 that the long-term safety of nuclear reactors designed by B&W be reassessed.
Additionally, the NRC staff concluded that because of the improvements already made to the plants, continued operation of the B&W plants during the reassess-ment would not pose an undue risk to public health and safety.
The objective of the overall assessment is to better understand the operational l
l characteristics of B&W plants and to compare the overall safety of these plants with other pressurized water reactors.
The approach used includes, among other things, (1) the review of operating experience at B&W plants, including the outcome of the Davis-Besse and Rancho Seco Incident Investigation Team reports; (2) a review of human factors aspects, including procedure reviews and inter-views of operation and maintenance personnel; (3) an examination of systems, such as the integrated control system and feedwater systems, that contribute to the complexity of pre-trip and post-trip reactor behavior; and (4) the perfor-mance of sensitivity and risk analysis.
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The Honorable Sam Nunn Since the reassessment began, more than 170 recommendations have been referred to the B&W licensees, some of which have been or are now being implemented.
These changes have already had a positive effect on safety by bringing about improvements; for example, improvements in the integrated control system and in the performance of the main feedwater system have reduced challenges to safety systems caused by feedwater transients.
Because of these and other changes, the staff has concluded that the plants are safer now than they were a year ago.
Further, implementation of the pertinent recommendations being proposed should result in additional improvements in plant performance and reductions in the frequency and severity of plant trips and transients.
Considerable work has already been completed by the NRC staff.
The NRC staff expects to complete its reassessment of B&W nuclear reactors by October 31, 1987.
The results of the staff's review will be documented in a safety evalua-tion report.
The fourth statement Dr. Gross made was:
"NRC says it does not have confidence in the computer model used to predict the chain of events that could be triggered by a severe accident at a B&W plant.
The reason is that the computer model was developed for plants designed by Westinghouse not the unique design used by B&W."
The " computer model" referred to by Dr. Gross is the NRC staff's "best estimate computer code," which is designed to accurately predict complex transients in B&W plants.
He has apparently referenced NRC staff recommendations regarding this "best estimate code" which were made several years ago, before integral systems test facilities applicable to the B&W design were available.
Although some experimental data from integral systems tests applicable to the B&W plant design have now been obtained, the NRC staf f still believes that additional data are required to enhance the accuracy of this code.
The implication of Dr. Gross' statement is that the NRC staff's code referred to by Dr. Gross is the same as the conservative computer code used in the plant safety analysis described in the Final Safety Analysis Reports (FSAR).
This is not the case.
The assumptions for the codes used in the FSARs were chosen to
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maximize the consequences of a potential accident--such as peak pressure, departure from nucleate boiling, and peak cladding temperature--for a range of design-basis events.
Thus, while the gathering of additional experimental data will enhance the NRC's ability to better understand and predict B&W plant i
behavior during complex transients, the current B&W plant safety analyses are I
based on the conservative codes and are sufficient to conclude that continued plant operation poses no undue risk to the public health and safety.
The fif th statement Dr. Grosn maus was:
"The safety research needed to provide the NRC with the necessary basic understanding of B&W plant behavior during accidents has 1
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uh The Honorable Sam Nunn been indefinitely postponed because of cuts in NRC's research budget and the refusal of B&W and the utilities to fund such research."
Again, Dr. Gross is referring to the NRC's efforts to obtain additional experi-mental data to improve the accuracy of the staff's "best estimate computer code." As stated earlier, this effort is not related to the plant safety analyses that were the basis for licensing of the B&W plants.
Some of the previously planned research to enhance the NRC staff's understanding of B&W plant behavior has been indefinitely postponed for the reasons Dr. Gross 1
I stated.
However, much research has already been completed.
Notably, 51 tests have been performed to date using the Multi-Loop Integral Systems Test (MIST) facility.
The NRC is continuing to actively seek industry support for additional research.
In addressing the other issues in his letter, Dr. Gross stated that the NRC q
has said that there is a 10% chance of a major accident in the next 20 years at one of the seven B&W plants currently operating, three of which are on the Savannah River.
From this he extrapolates that there is a 1 in 200 chance that a home near one of these plants will become unlivable next year, and he terms this risk unacceptable.
Although the exact source of the 10% to 50% probability statement is unclear, it appears to be consistent with previously published NRC staff estimates of core meltdown frequency for B&W plants.
However, it must be pointed out that the occurrence of severe core damage does not necessarily mean that significant releases of radioactivity would occur.
The probability of significant releases of radioactivity is less than the probability of a core meltdown.
Dr. Gross has apparently equated the term " severe accident" to " core meltdown combined with containment failure."
In general, the use of probability risk assessments (PRAs) for a number of plants has been beneficial in estimating an average accident probability.
PRAs have also been useful in identifying those plants with probabilities that are significantly different from the average.
PRA's i
also identify the individual contributors to risk at a specific plant.
Added attention can then be focused on addressing the principal contributors to risk at that plant.
Care must be used when drawing conclusions regarding a specific plant or small group of plants because plant-specific design and procedure differences may have a substantial effect on the results.
Finally, with regard to the statement by Dr. Gross that ample sources of power exist in areas served by the B&W plants, to the extent it is implied that the NRC would allow the need for electrical power in the area served by a utility licensee to become a dominant consideration in NRC decisions regarding the safe operation of nuclear power plants, let me assure you this is not the case.
The NRC does not actively monitor the power generating sources or power requirements
.The Honorable Sam Nunn 1 of its licensees.
Should safety problems be identified at'any or all of the.
B&W plants, the NRC would take appropriate action to protect the public~ health and safety.
I trust that this information will assist you in responding to Dr. Gross.
Sincerely, M
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V. SteUC l
Victor Stello, Jr.
I Executive Director for Operations i
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being satisfactorily addressed by the licensee, Duke Power Company.
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these plants not meet NRC regulations for any reason, the NRC will take appro-q priate action.
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Next, Dr. Gross stated that the NRC has said that there is a 10-50% chance of j
a major accident in the next 20 years to one of the 7 B&W plants currently l
operating. He concluded that a home would have a 1/200 chance of becoming i
unlivable next year. Although the exact source of the above probability statement is unclear, it appears to be consistent with previously published NRC 1
staff estimates of core meltdown frequency for B&W pl' ants.
However, it must be pointed out that the occurrence of severe core damage'does not necessarily mean i
that significant releases of radioactivity would occur ( The probability of that occurring is much less than the probability of a core meltdown. Dr. Gross has apparently equated the term " severe accident" to core meltdown combined with containment failure. So, in general one must be ca'reful when drawing conclusions from reactor accident probability estimates. \\ Also it is difficult to estimate with precision what the actual industry-wide 1ikelihood of a severe core damage accident is today, since probability risk asse,ssments (PRAs) have not been done on all plants. Plant-specific design and procedure differences can make substantial differences in PRA results; and the reiults themselves have substantial uncertainties.
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Dr. Gross implied that the NRC would allow the need for electrical power in the area served by a utility licensee become a dominant consideration when making decisions regarding the safe operation of nuclear power plants \\.
This is not Infact,theNRCdoesnotactivelymonitorthepower\\problemsthat the case.
generating sources or power requirements of its licensees.
Should safety would require shutdown be identified with any or all of the B&Wiplants, public health and safety would be the foremost concern.
Power requirements would only be a minor consideration if considered at all.
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Dr. Gross expressed particular concern about the three 8&W designed plants at Oconee on the Savannah River. These plants currently operate under; licenses issued by the NRC, and NRC concerns applicable to operating power reactors are being satisfactorily addressed by the licensee, Duke Power Company.\\ Should these plants not meet NRC regulations for any reason, the NRC will take appro-priate action.
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I trust that this information will assist you in responding to your constituent.
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i Sincerely, Victor Stello, Jr.
Executive Director for Operations p
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'FROM:
DUE: 08/31/87 EDO. CONTROL: 003103 DOC DT 08/03/87; SEN.' SAM NUNN FINAL REPLY:
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OCA FOR SIGNATURE OF GREEN
-SECY NO: 87-976 i
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ROUTINO:
ENCLOSES LETTER FROM THOMAS F. GROSS CONCERNING NURCLEAR POWER PLANTS DESIGNED BY B&W i
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SPECT.A.. INSTRUCTIONS OR REMARKS:
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OFFICE OF THE SECRETARY CORRESPONDENCE CONTROL TICKET PAPER NUMBER:
CRC-87-0976 LOGGING DATE: Aug 14 87 ACTION OFFICE:
EDO AUTHOR:
S. Nunn--Const Ref AFFILIATION:
U.S. SENATE LETTER DATE:
Aug 3 87 FILE CODE: ID&R-5 B&W
SUBJECT:
Nuclear. power plants designed by the B&W company ACTION:
Direct Reply DISTRIBUTION:
OCA to Ack SPECIAL HANDLING: None NOTES:
Thomas Gross DATE DUE:
Aug 28 87 SIGNATURE:
DATE SIGNED:
AFFILIATION:
Rec'd Dil. ED3 Date S-/7-89 Time.
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